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Wilson, Paul P. H.

Publications and source records attributed to Wilson, Paul P. H..

11 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Hardware-Accelerated Ray Tracing of CAD-Based Geometry for Monte Carlo Radiation Transport

Monte Carlo radiation transport (MCRT) methods have been used to simulate radiation environments for many decades by tracking individual particles through a model to accumulate statistical information. MCRT geometry is historically formed using the constructive solid geometry (CSG). Recently, significant work has been performed to support simulations using computer-aided design (CAD)-based tessellated surfaces to support highly complex geometries. Ray tracing acceleration data structures from the rendering and visualization community are applied to accelerate particle tracking in CAD-based models. Despite these efforts, CSG representations provide the superior performance in surface intersection operations during particle flight. Concurrently, pseudo Monte Carlo methods have become prevalent in rendering applications to support more realistic models for scattering media, motivating innovations that are advantageous for MCRT simulations. Finally, the authors’ work extends these innovations by employing Intel’s Embree ray tracing kernel within a geometry toolkit for Monte Carlo to improve the simulation performance using CAD-based models by factors of 1.5 to 2.

42 ENGINEERING↗
DOE OSTI2017

Particle tracking acceleration via signed distance fields in direct-accelerated geometry Monte Carlo

Computer-aided design (CAD)-based Monte Carlo radiation transport is of value to the nuclear engineering community for its ability to conduct transport on high-fidelity models of nuclear systems, but it is more computationally expensive than native geometry representations. This work describes the adaptation of a rendering data structure, the signed distance field, as a geometric query tool for accelerating CAD-based transport in the direct-accelerated geometry Monte Carlo toolkit. Demonstrations of its effectiveness are shown for several problems. The beginnings of a predictive model for the data structure's utilization based on various problem parameters is also introduced.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Recovering Topology of Nested Volumes Represented by Single Closed Surfaces

As scientific computing abilities continue to advance, so too have the efforts to model and simulate highly complex geometries in radiation environments. Modeling has advanced from primitive constructive solid geometry (CSG) to voxelized models, and now to high fidelity polygon mesh models. While polygon mesh surface models are very accurate, most Monte Carlo codes do not allow for their direct use. Often they are converted to voxel or tetrahedral mesh models and then used with the physics codes. The Direct Accelerated Geometry Monte Carlo (DAGMC) software package has the ability to efficiently run transport calculations directly on tessellated surface models. Typically the surface meshes used for DAGMC calculations come from CAD-based geometry models generated by solid modeling software. In some cases geometries of interest, like human phantoms, are generated by other means and may not contain all of the necessary information for particle transport. For example, mesh geometries in the OBJ file format contain vertex and connectivity data for each surface, but do not have any information about surface topology. These types of file formats necessitate a tool that can build a hierarchy of surfaces and reconstruct the topological information. The generate hierarchy tool was developed for this purpose. The following summary will explain the algorithm behind the tool, describe a test case, and finally demonstrate a real-world application of the tool with an MCNP5 calculation that directly uses a polygon mesh human phantom.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

University of Wisconsin Nuclear Reactor Modeling Improvements

The University of Wisconsin Nuclear Reactor converted from HEU to LEU fuel in 2009. As part of the conversion process, an MCNP model was developed to predict flux distributions and depletion in the core. This core model included an unforeseen excess reactivity bias, resulting in an overestimate of core reactivity, and requiring the core configuration to be shuffled in 2011. The original model was constructed by hand and in order to account for radial and axial burnup in the fuel, and to include sufficient detail in the geometric features, the input file was over 17,000 lines long. Given the challenges in assessing the quality of such a long and complicated MCNP model, this project was designed to build an MCNP model based on a minimal set of input quantities in a modular and repeatable fashion. A Python code was implemented to write the MCNP input file from fundamental reactor data, both geometric and material. The recent conversion to LEU fuel provides a unique opportunity to characterize fuel composition quite accurately in the core. The goal of this project is to produce an accurate, maintainable model that can be used with future reactor physics experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 10 4 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

A spectral analysis of the domain decomposed Monte Carlo method for linear systems

The domain decomposed behavior of the adjoint Neumann-Ulam Monte Carlo method for solving linear systems is analyzed using the spectral properties of the linear oper- ator. Relationships for the average length of the adjoint random walks, a measure of convergence speed and serial performance, are made with respect to the eigenvalues of the linear operator. In addition, relationships for the effective optical thickness of a domain in the decomposition are presented based on the spectral analysis and diffusion theory. Using the effective optical thickness, the Wigner rational approxi- mation and the mean chord approximation are applied to estimate the leakage frac- tion of random walks from a domain in the decomposition as a measure of parallel performance and potential communication costs. The one-speed, two-dimensional neutron diffusion equation is used as a model problem in numerical experiments to test the models for symmetric operators with spectral qualities similar to light water reactor problems. We find, in general, the derived approximations show good agreement with random walk lengths and leakage fractions computed by the numerical experiments.

97 MATHEMATICS AND COMPUTING↗